This paper investigates the effect of seashell (SS) powders of 75 μm size, reinforced in multi-walled carbon nanotube (MWCNT)-added nylon 66 polymer composite, by studying its mechanical and thermal properties for suitable application in the automotive industry. Seashells were collected from the seashores and the mechanical ball milling method is employed to ground the SSs into powders using a sieve machine, and the desired particulate size of 75 μm is obtained. These particulate SSs are reinforced in the matrix of nylon 66 in different proportions, viz., 3, 6, 9, 12, and 15% by weight along with 1 wt.% of MWCNTs. Twin-screw extruder and injection molding are employed to prepare the testing specimens. Mechanical and thermal tests were carried out as per ASTM standards. From the mechanical and thermal analyses, it is observed that tensile strength and coefficient of linear thermal expansion decrease with an increase in SS content, whereas flexural strength, shore hardness, impact strength, Vicat softening point, and heat deflection temperature rise with higher wt.% of SSs. This is owing to the thermoplastic’s decreased flexibility, which increases its resilience to deflection and load and its high thermal stability. The added MWCNTs improved the bonding strength and mechanical properties due to their higher surface areato- volume ratio.

최근 전기차 시장의 확장으로 배터리 산업이 급격히 성장함에 따라 폐배터리 리사이클링 기술 개발의 필요성이 증가하고 있다. 폐배터리 리사이클링 기술은 배터리 산업에 핵심적인 리튬, 코발트, 니켈 등 희소금속의 공급을 안정화하고 환경 및 인간의 건강에 미치는 영향을 경감할 수 있다. 본 총설에서는 금속 회수 기술의 배경이 되는 이론적 원리와 현재 상 용되고 있는 금속 회수 공정을 소개하고자 한다. 또한, 기존 공정의 문제점을 개선하려는 연구 및 기술 개발 동향을 서술하여 리사이클링 기술이 나아가야 할 방향을 소개하고자 한다.

Recently, BNS (Best System) developed a system for evaluation and classification of soil and concrete wastes generated from nuclear power plant decommissioning. It is composed of various modules for container loading, weight measurement, contamination evaluation, waste classification, stacking, storage and control. The contamination evaluation module of the system has two sub modules. One is for quick measurement with NaI (Tl) detector and the other is for accurate measurement with HPGe detector. The container used at the system for wastes handling has capacity of 100 kg and made of stainless steel. According to the measurement result of Co-60 and Cs-137, the waste is classified as waste for disposal or waste for clearance. Performance of the system was demonstrated using RM (Reference Material) radiation source. This year, necessity of system improvement was suggested due to revised operation requirements. So, the system should show throughput of more than 1 ton/hr and Minimum Detectable Activity (MDA) of less than 0.01 Bq/g (1/10 of criteria for regulatory clearance) for Co-60 and Cs-137. And soil waste become main target of the system. For this, the container used for soil waste handling should have capacity of 200 kg. As a result, material for the container need to be changed from stainless steel to plastic or FRP (Fiber Reinforced Plastics). And large area detector should be introduced to the system to enhance processing speed of the system. Additionally, container storage rack and conveyor system should be modified to handle 200 kg capacity container. Finally, moving path of the container will be redesigned for enhanced throughput of the system. In this paper, concept development of the system was suggested and based on that, system development will be followed.

The immobilization of low- and intermediate-level radioactive waste (LILW) is crucial for its final disposal in repositories. While cementitious waste forms have conventionally been used for immobilizing various LILWs, they suffer from several issues, including poor durability, low resistance to leaching, and limited waste loading capacity. As an alternative, alkali or acid-activated geopolymer waste forms have garnered global attention. Unlike cementitious waste forms, geopolymer waste forms exhibit excellent physicochemical characteristics due to their three-dimensional amorphous structure and low calcium content. In this work, we provide an overview of geopolymer waste form research being conducted in countries such as Japan, the United Kingdom, the European Union, and South Korea. We specifically focus on the immobilization of soil waste, spent ion exchange resins, organic liquid waste, and evaporator concentrate (borate waste). We also identify the factors influencing the physicochemical characteristics of geopolymer waste forms and their immobilization performance. We propose a guide for optimizing the molar mixing formulations of geopolymer waste forms, including the selection of appropriate precursor materials. Additionally, we discuss the future prospects and significant challenges in the field of geopolymer waste forms that need to be addressed for their application in radioactive waste management.

In the decommissioning site of Korean Research Reactor 1&2 (KRR-1&2), according to Low and Intermediate-level Radioactive Waste Disposal Acceptance Criteria of the Korea Radioactive Waste Agency (WAC-SIL-2022-1), characteristics of radioactive waste was conducted on approximately 550 drums of concrete and soil waste for a year starting from 2021. Among them, 50 drums of concrete waste transported and disposed to Gyeongju LILW disposal facility at the end of 2022. For the remaining approximately 500 drums of concrete and soil waste stored on-site, they were reclassified into two categories: permanent disposal grade and clearance grade. This classification was based on calculating the sum of fractions (SOF) per drum for each radionuclides. The plan is to dispose of around 200 drums in the permanent disposal grade and about 300 drums in the clearance grade by the end of 2023. Since concrete and soil decommissioning wastes are generated in large quantities over a short period with similar origins, they were grouped within five drums as suggested by the acceptance criteria. Mixed samples were collected from each group and used for radionuclide analysis. When utilizing mixed samples, three distinct samples are collected and analyzed for each group. The maximum value among these three radionuclide analysis results is then uniformly applied as the radionuclide concentration value for all drums within that group. Radioactive nuclides contained in similar types of radioactive waste with similar origins can be expected to have some statistical distribution. However, There has been no verification as to whether the maximum value among the three mixed samples exists within the statistical distribution or if it deviates from this distribution to represent a different value. In this study, we confirmed characteristics of radionuclide concentration distribution by examining and comparing radionuclide concentration distributions for radioactive wastes drum grouped for nuclear characteristic among 50 concrete wastes drum disposed in year 2022 and 500 concretes & soils drum scheduled for disposal (clearance or permanent disposal) in year 2023. In particular, when comparing tritium to other nuclides, it was observed that the standard deviation for the distribution of maximum values was approximately 318 times larger.

Various radioactive metal wastes are generated during operation and decommissioning of nuclear facilities. Radioactive metal wastes with complex geometries or volumetric contamination can be difficult to decontaminate and disposal costs may increase. To solve these problems, the radioactive metal wastes can be treated by melting method. In this study, we designed a melting furnace system of air induction melting type, which is widely utilized due to its advantages of good thermal efficiency, uniform heating and guaranteed safety for radioactive material. By utilizing the melting furnace system, volatile radionuclides existed in the base material can be captured in the form of gas or dust by the filter. The radionuclides whose chemical properties can easily form metal oxides present as slag. For this reason, the specific radioactivity of the base material can be reduced. Radionuclides that are difficult to transport to slag and dust are uniformly distributed in the base material. A dedicated power supply and a transformer were necessary to be included in the melting furnace system since the induction furnace uses high-frequency currents. In addition, a hood is placed on top of the furnace to capture fumes generated during melting, and additional hoods were installed around the furnace to remove airborne dust. In particular, a dust collection unit consisting of a cyclone and a HEPA filter were constructed to effectively collect dust containing radionuclides. During the melting process, the slag is removed and accumulated separately, and the ingot production system was designed to produce the ingot using molten metal. The furnace was constructed for tilting the molten metal by moving the furnace using hydraulic system. The water cooling system and cooling tower were prepared to cool off the equipment with high temperature during melting is cooled off. The above process was specified in the operating procedure developed for this melting furnace system, and the operator shall operate and inspect according to the prescribed procedures. The radioactivity concentration in the sample taken in the step of tilting shall be analyzed whether they meet clearance level for self-disposal determined and publicly announced by the Commission. We can conduct self-disposal for the product of melting furnace system confirmed by the Commission as having the radioactivity concentration by nuclide not exceeding the value determined by the Commission.

Recently, the nuclear decommissioning and environmental restoration industries has significantly attracted as a new industry field due to the decision to decommission the KORI#1 and WOLSONG #1 nuclear power plant. In order to dispose of the decommissioning radioactive wastes generated during nuclear decommissioning, proper analysis is required, and disposal decisions are determined based on the analysis results. When dismantling a nuclear power plant, a few thousand of tons decommissioning waste are produced, so these require analysis for proper disposal. Therefore, a radionuclide facility for decommissioning waste analysis is essential for the disposal of the large quantities of decommissioning waste generated during nuclear power plant decommissioning. Korea Research Institute of Decommissioning (KRID) was established radionuclide analysis facilities to address above issues and support nuclear power plant decommissioning projects. The plan is to perform classification by type and radionuclide for all waste produced during nuclear power plant decommissioning and to support the disposal of radioactive wastes. In addition, we plan to establish validation methods for samples where verification methods are not established, in order to conduct efficient analysis and management. In this presentation, we will introduce the radionuclide facility currently under construction at KRID and present the space design, equipment layout, and utilization plans.

Structural stability of a waste form can be provided by the waste form itself (steel components, etc.), by processing the waste to a stable form (solidification, etc.), or by emplacing the waste in a container or structure that provides stability (HICs or engineered structure, etc.). The waste or container should be resistant to degradation caused by radiation effects. In accordance with the requirements for the domestic waste acceptance criteria, irradiation testing of solidified waste forms containing spent resin should be conducted on specimens exposed to a dose of 1.0E+6 Gy and other material 1.0E+7 Gy. Expected cumulative dose over 300 years is about 1.770E+6 Gy for spent resin and 0.770E+6 Gy for dried concentrated waste generated from NPPs generally. According to NRC Waste Form Technical Position, to ensure that spent resins will not undergo adverse degradation effects from radiation, resins should not be generated having loadings that will produce greater than 1E+6 Gy total accumulated dose. If it necessary to load resins higher than 1E+6 Gy, it should be demonstrated that the resin will not undergo radiation degradation at the proposed higher loading. This is the recommended maximum activity level for organic resins based on evidence that while a measurable amount of damage to the resin will occur at 1E+6 Gy, the amount of damage will have negligible effect on disposal site safety. Cementitious materials are not affected by gamma radiation to in excess of 1E+6 Gy. Therefore, for cement-stabilized waste forms, irradiation qualification testing need not be conducted unless the waste forms contain spent resins or other organic media or the expected cumulative dose on waste forms containing other materials is greater than 1E+7 Gy. Testing should be performed on specimens exposed to IE+6 Gy or the expected maximum dose greater than 1E+6 Gy for waste forms that contain ion exchange resins or other organic media or the expected maximum dose greater than 1E+7 Gy for other waste forms. This is suggestion as a review result that requirement for irradiation testing of solidified waste forms has something to be revise in detail and definitively.

This study aimed to provide better understanding of the bedrock aquifer bacterial communities and their functions in deep geological repository (DGR) environment. Two study sites of uranium deposits in the Ogcheon Metamorphic Belt were selected: Boeun and Guemsan. From two study sites, six groundwater samples were obtained with different boreholes and depths: OB1 (Boeun, 25 m), OB3 (Boeun, 80 m), GS1 (Guemsan, 25 m), GS2 (Guemsan, 85-90 m), GS3-I (Guemsan, 32- 38 m), GS3-II (Guemsan, 70-74 m). The physicochemical properties of groundwater were analyzed by multi-parameter sensors, ion chromatography (IC), and inductively coupled plasma optical emission spectroscopy (ICP-OES). Illumina Miseq sequencing was performed to investigate bacterial community in six groundwater samples. In addition, the number of sulfate-reducing bacteria (SRB) was quantified by a quantitative PCR (qPCR). Bacterial community composition varied in response to boreholes and depths. A total of 14 different phyla and 36 classes were detected from six groundwater samples. Overall, Proteobacteria, Actinomycetota, and Bacteroidota were dominant in the phylum level. SRB and iron-reducing bacteria (IRB) were detected in all groundwater samples even though organic carbon sources were not abundant (0.7-3.3 mg-total organic carbon/L). This result shows a potential to immobilize uranium in DGR environment. In particular, SRB, Desulfosporosinus fructosivorans and Humidesulfovibrio mexicanus were mainly detected in GS1 and GS2 groundwater samples, which attributed to higher dissimilatory sulfite reductase functional gene copy number in GS1 and GS2 groundwater samples. Statistical analysis was performed to understand the correlation between environmental factors and core bacterial species. Dissolved oxygen (DO), Fe, and Mn concentrations were positively correlated with Curvibacter fontanus while Undibacterium rivi had a negative correlation with pH. These results indicate that bacterial community could be changed in response to environmental variation. Further study with a greater number of samples is necessary to obtain statistically reliable and meaningful results for a safe DGR system.

The use of NG in the transportation sector is becoming an appealing option to diesel and gasoline fuels, presenting higher benefits. ANG technology offers a secure, cost-effective, energy-efficient strategy for the storage of NG in porous sorbents at reasonable gas densities. The major goal for its extensive utilization is the requirement of effective storage materials under practicable conditions. Recently, there has been increased attention in utilizing bio-wastes for the preparation of microporous carbons. In this contribution, our growing knowledge on the use of biobased materials and the processing strategies in an effort to predictively produce effective porous carbons appropriate for ANG technology have been reviewed. By careful literature selection, different precursors with different alternative processes to convert low-cost bio-wastes into porous carbons and achievements in methane storage are presented. To gain deeper insight into the technology, the correlation between the structural and chemical properties of materials and the factors affecting the storage performance are highlighted. The utilization of bio-wastes for the development of microporous carbons with facile methods emerged to be encouraging, which would be significant in larger scale applications. Bio-waste processing for ANG storage is valued over many other techniques, and the products are able to store substantial levels of methane. This review could help improve researchers’ evaluation of the methods as a guideline for ANG. Further studies for achieving an accomplished interconnection between the structural characteristics and the methane storage capacities with different bio-wastes and optimization strategies would be beneficial.

To analyze the radioactivity of 3H and 14C in miscellaneous radioactive wastes generated from nuclear power plants, a wet digestion method using sulfuric acid is currently used. However, sulfuric acid is classified as a special management material, and there is no disposal method for contaminated radioactive waste. Therefore, research on a thermal decomposition method that can analyze the DAW radioactive waste samples without using sulfuric acid is necessary. In this study, we will cover the final sample amount, sample injection method, and prevention of organic ignition to meet the minimum detection limit requirements of the analysis equipment. Through this research, optimal conditions for the thermal decomposition method for analyzing the radioactivity of 3H and 14C in DAW radioactive wastes generated from nuclear power plants can be derived.

Support for nuclear power plant (NPP) dismantling & decommissioing (D&D) industry is necessary through development of the infrastructure and the D&D technology. Because KORI#1 and Wolsong#1 is planned to decommission until around 2030. Korea research institute of decommissioing (KRID) was established through the preliminary feasibility study. KRID has plan to support nuclear companies to join D&D industry. Normal facilities (Lv.1) of KRID infracstucture are currently being constructed and radiation management facilities (Lv.2) construction is expected to begin in October. Further, KRID is planning the construction of equipment to develop the procedure for radionuclide analysis through R&D project. A total period of the R&D project is 45 months, and the total R&D funding for this period is 19.4 billion won. The ultimate goal of the R&D project is to build the infractstucture base to analyze decommissioning radioactive wastes. Furthermore, the R&D project is important to reliably perform the NPP D&D.

Low- and intermediate-level radioactive waste for permanent disposal often contains organic complexing agents, so-called chelating agents. Organic complexing agents, which are polycarboxylic acids, can increase the mobility of radionuclides into the environment by forming water-soluble complexes with most heavy metals. Therefore, analyzing the complexing agents in radioactive waste is crucial for comprehensive management of nuclear wastes. According to regulatory guidelines, specifically Notice No. 2021-16 issued by the Nuclear Safety and Security Commission, the determination of chelating agent content in radioactive waste materials is required to ensure proper management and safe disposal. However, only a few methods are available to analyze the chelators in various matrices such as concrete, metals, soil, and mixed solid wastes like plastics, vinyl, and rubber. Recently, we found a UV-Vis method based on an enzymatic reaction is inadequate for analyzing citric acid in radioactive waste with a complex matrix like concrete. To address this, we developed a method to determine the contents of EDTA and NTA using a UV-Vis spectrophotometer and citric acid using ion chromatography. The results showed good validity and reliability to determine the chelating agents in various radioactive wastes.

During decommissioning and site remediation of nuclear power plant, large amount of wastes (including radioactive waste) with various type will be generated within very short time. Among those wastes, soil and concrete wastes is known to account for more than 70% of total waste generated. So, efficient management of these wastes is very essential for effective NPP decommissioning. Recently, BNS (Best System) developed a system for evaluation and classification of soil and concrete wastes from the generation. The system is composed of various modules for container loading, weight measurement, contamination evaluation, waste classification, stacking, storage and control. By adopting modular type, the system is good for dealing with variable situation where system capacity needs to be expanded or contracted depending on the decommissioning schedule, good for minimizing secondary waste generated during maintenance of failed part and also good for disassemble, transfer and assemble. The contamination evaluation module of the system has two sub module. One is for quick measurement with NaI(Tl) detector and the other is for accurate measurement with HPGe detector. For waste transfer, the system adopts LTS (Linear Transfer System) conveyor system showing low vibration and noise during operation. This will be helpful for minimizing scattering of dust from the waste container. And for real time positioning of waste container, wireless tag was adopted. The tag also used for information management of waste history from the generation. Once a container with about 100 kg of soil or concrete is loaded, it is moved to the weight measurement module and then it transfers to quick measurement module. When measured value for radioactivity concentration of Co- 60 and Cs-137 is more than 1.0 Bq/g, then the container is classified as waste for disposal and directly transferred to stacking and storage rack. Otherwise, the container is transferred to accurate measurement module. At the accurate module, the container is classified as waste for disposal or waste for regulatory clearance depending on the measurement result of 0.1 Bq/g. As the storage rack has a sections for disposal and regulatory clearance respectively, the classified containers will be positioned at one of the sections depending on the results from the contamination evaluation module. The system can control the movement of lots of container at the same time. So, the system will be helpful for the effective nuclear power plant decommissioning in view of time and budget.

Disposal of radioactive waste requires radiological characterization. Carbon-14 (C-14) is a volatile radionuclide with a long half-life, and it is one of the important radionuclides in a radioactive waste management. For the accurate liquid scintillation counter (LSC) analysis of a pure beta-emitting C-14, it should be separated from other beta emitters after extracted from the radioactive wastes since the LSC spectrum signals from C-14 overlaps with those from other beta-emitting nuclides in the extracted solutions. There have been three representative separation methods for the analysis of volatile C-14 such as acid digestion, wet oxidation, and pyrolysis. Each method has its own pros and cons. For example, the acid digestion method is easily accessible, but it involves the use of strong acids and generates large amount of secondary wastes. Moreover, it requires additional time-consuming purification steps and the skillful operators. In this study, more efficient and environment-friendly C-14 analysis method was suggested by adopting the photochemical reactions via in-situ decomposition using UV light source. As an initial step for the demonstration of the feasibility of the proposed method, instead of using radioactive C-14 standards, non-radioactive inorganic and organic standards were investigated to evaluate the recovery of carbon as a preliminary study. These standards were oxidized with chemical oxidants such as H2O2 or K2S2O8 under UV irradiations, and the generated CO2 was collected in Carbo-Sorb E solution. Recovery yield of carbon was measured based on the gravimetric method. As an advanced oxidation process, our photocatalytic oxidation will be promising as a time-saving method with less secondary wastes for the quantitative C-14 analysis in low-level radioactive wastes.

With the rapid growth of nuclear power in China, a large number of dry wastes, which mainly include the high efficiency particulate air filters (glass fiber), cotton, polyethylene, and absorbent paper with low-level radioactivity and high volume, will be produced during the operation and maintenance of the nuclear power plants. Thermal plasma treatment is a world acceptable technology to incinerate and immobilize radioactive wastes, owing to the high volume reduction factor and the excellent chemical durability of the vitrified waste form. China has developed thermal plasma technology for the treatment of dry wastes from nuclear power plants for more than 15 years and the pilot plant has been constructed. This work will concentrate on the formulation of waste glass fiber to adapt to the vitrification process. A three-component (glass fiber-CaO-Na2O) constrained-region mixture experiment was designed and their viscosity data was mainly studied. The quadratic Scheffé model was used to plot the component effect on melting temperature. The retentions of simulated nuclides, such as Co, Sr, and Cs in the glasses were analyzed. In addition, the glass fiber as a glass matrix to immobilize residual ashes from the thermal plasma gasification of cotton, polyethylene, and absorbent paper was investigated as well.